• 한국농공학회지
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• 제38권3호
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• pp.82-91
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• 1996

This is a basic study for the static and dynamic analysis on the elasto-plastic and elasto-viscoplastic of an axi-symmetric shell. The objective of this study was to investigate the mechanical characteristics of a nuclear reinforced concrete containment structure, which was selected as a model, by a numerical analysis using a finite element method. The structure was modeled with discrete ring elements of 8-noded isoparametric element rotating against the symmetrical axis, and the interaction between the foundation and the structure was modeled by Winkler's model. Also, the meridional tendon was modeled with 2-node truss elements, and the hoop tendon was done with point elements in two degrees of freedom. The effect of the tendon was considered without the increasement in total degree of freedom as the stiffness matrix of modeled tendon elements was assembled on the stiffness matrix of ring elements linked with the tendon. The results obtained from the analysis of an example were summarized as follows : 1. The stresses in the hoop direction on the interior and exterior surfaces of the structure were shown in changes of similar trend, and high stresses appeared on the structure wall 2. The stresses in the meridional direction on the interior and exterior surfaces were shown in change of different trend. Especially, the stresses at the junctions between the dome and the wall and between the wall and the bottom plate of the structure were very high, compared with those at other parts of the structure. 3. The stress changes in the direction of thickness on the crown of the dome were much linearly distributed. However, as the amount of tendon increased, the stresses in the upper and lower parts of the wall established with the tendon were shown stress concentration. 4. The stress changes in the direction of thickness on the center of the structure wall was linearly distributed in the all cases, and special stress due to the use of the tendon was not shown.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.199-206
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• 1997

For the Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside its containment to achieve cost and safety Improvement. To apply LBB concept to MSL, leak sensors highly sensitive to humidity is required. In this paper, a ceramic material, MgCr$_2$O$_4$-TiO$_2$ has been developed as a humidity sensor for MSL applications. Experiments peformed to characterize the electrical conductivity shows that the conductivity of MgCr$_2$O$_4$-TiO$_2$ responds sensitively to both temperature and humidity changes. At a constant temperature below 10$0^{\circ}C$, the conductivity increases as the relative humidity increases, which makes the sensor favorable for application to the outside of MSL insulation layer But as temperature increases beyond 10$0^{\circ}C$, the sensor composition should be adjusted for the application to KNGR is to be made at temperature above 10$0^{\circ}C$.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.907-914
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• 2015

The energetic steam explosion caused by contact between the high temperature molten core and water is one of the phenomena that may threaten the integrity of the containment vessel during severe accidents of light water reactors (LWRs). We examined the dependence of steam explosion loads in a typical reactor cavity geometry on selected model parameters and initial/boundary conditions by using a steam explosion simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters, we put an emphasis on the water pool depth that has significance in terms of accident mitigation strategies including cavity flooding. The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load. The water pool depth also showed a significant impact. The energy conversion ratio based on the enthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, we proposed a simplified method for evaluation of the steam explosion load. The results showed fair agreement with JASMINE.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• 방사성폐기물학회지
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• 제20권4호
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• pp.411-428
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• 2022

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2079-2091
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• 2024

In this paper, a model is developed for calculating pH in the presence of organic impurities due to dissolution of paint and/or continuous injection of organic impurities in the sump. The model is implemented in the AnCheBi code for the analysis of chemical behaviors of the iodine in the containment when the pH changes during a severe accident. Validation of the model is performed with P10T2 and P11T1 experiments carried out by AECL in Canada under the BIP project. Importance analyses of the pH calculation model in the AnCheBi code are then performed with the aforementioned experimental data via Latin hypercube sampling on the reaction coefficients, sensitivity analyses of AnCheBi, and calculation of the correlation coefficients between the reaction coefficients and figure of merits (the pH and the concentrations of the various iodine species). From the importance analyses, we provide the sensitivity of the pH calculation model to the change of pH and the concentrations of the various iodine species and the reaction coefficients related with the dominant phenomena underlying the change of pH and the concentrations of the species.

• 한국전산구조공학회논문집
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• 제24권6호
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• pp.715-722
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• 2011

본 논문에서는 항공기 충돌에 의한 원전 격납건물의 거동을 병렬해석을 통해 수행하였다. 지금까지의 원전 격납건물에 대한 항공기 충돌관련 연구는 항공기의 경우, Riera의 충격하중-시간함수를 이상화하여 대상 구조체의 일정영역에 대해 충격하중으로 적용하는 방법을 사용해 왔고 충돌대상 구조체의 경우, 단순 철근콘크리트 벽체나 빌딩에 머물러 왔다. 하지만 본 논문에서는 항공기(Boeing-767, http://www.boeing.com)와 가상의 원전 격납건물을 실제와 유사하게 모델링하여 해석을 수행하였으며, 항공기모델은 충돌평가 가이드인 NEI 07-13(2009)에서 허용하는 Riera의 식에 따른 충돌하중이력곡선과 비교하는 방법으로 검증되었다. 또한, 일반적으로 고속 충돌해석은 짧은 시간동안 두 개 이상의 물체가 접촉하고 동적 대변형을 일으키는 비선형성이 강한 문제로 많은 계산시간이 요구되기 때문에 이를 효과적으로 다루기 위해서는 단일 CPU만으로는 한계가 있다. 따라서 본 논문에서는 해석의 효율성을 향상시키기 위해 자체 구축한 리눅스 클러스터 시스템을 이용하여 Message-Passing MIMD 형태의 병렬해석을 수행하였고 병렬성능에 대한 평가를 위해 무근콘크리트(Plain Concrete, PC), 철근콘크리트(Reinforced Concrete, RC), 내부 Liner Plate를 부착한 철근콘크리트(RC with Containment Liner Plate, CLP), SC구조(Steel-Plate Concrete, SC)등 4가지 경우에 대한 수치해석 효율성이 비교 검토되었다.

• 한국안전학회지
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• 제23권2호
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• pp.21-29
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• 2008

To evaluate the structural integrity of the NPP containment building more rigorously, the effective prestress, which is one of the most affecting elements, needs to be estimated exactly. This paper presents the results of an experimental study to determine the effective prestress force in prestressed concrete beams. It is possible to improve the effective prestress measuring method by test beam, which is being applied for the investigation of the nuclear power plant in operation. If experimentally evaluated Lift-Off method in this study can be coupled with test beam test currently being used in in-service nuclear power plant, it is possible to measure prestress loss of the tendon and the level of the effective prestress load.